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At least 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Actinide Removal Process Modified Flowsheet Demonstration: Filter Only Operation - 17415

January 2016 marked the demonstration of the Actinide Removal Process (ARP) Filter-Only Flowsheet at the Savannah River Site's Liquid Waste Organization operated by Savannah River Remediation, LLC. This ongoing flowsheet demonstration has resulted in a marked increase in throughput over the historical Monosodium Titanate (MST) flowsheet through both increased facility availability and production rate. Not only has the Interim Salt Disposition Program been able to process salt waste more quickly, there has been more processing time between cleaning of the filter elements, which has the added benefit of reducing cesium-137 (Cs-137) loading to the Salt Disposal Facility (SDF). The new flowsheet is a promising improvement to salt waste processing, enhancing the mission of closing Waste Tanks at the Savannah River Site in Aiken, South Carolina. The demonstration of Filter-Only flowsheet has increased salt waste throughput, reduced Cs-137 load to the SDF, and may increase equipment longevity compared to the historical MST flowsheet. The gains in salt volume dispositioned through ARP/MCU benefits the safe, timely, and cost effective closure of legacy waste tanks at the Savannah River Site, reducing the most significant environmental risk in the State of South Carolina. (authors)

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DOE OSTI2018

Use of Robotics in Liquid Waste System - {sup 3}H Evaporator, Tank Closure and Defense Waste Processing Facility - 18525

The Savannah River Site (SRS) Liquid Waste System (LWS) safely stores and treats high-level radioactive waste. The LWS consists of 51 waste storage tanks (eight of which are operationally closed and filled with grout), waste evaporators, treatment facilities, and solidification facilities such as the Defense Waste Processing Facility (DWPF) and Salt-stone Production Facility (SPF). At Savannah River Remediation (SRR), the liquid waste contractor, robotic technologies are needed to enter areas and spaces that are otherwise inaccessible or prohibit direct access by workers due to unsafe, unstable, or unknown physical or structural conditions; or configurations that are hard to reach or beyond reach without taking extraordinary mechanical measures. As discussed below, SRR uses robotic technologies at various LWS facilities to reduce dose to the workers as well as provide access to unreachable locations due to the presence of radiological or chemical hazards. (author)

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DOE OSTI2015

Engineering Options Assessment Report: Nitrate Salt Waste Stream Processing

This report examines and assesses the available systems and facilities considered for carrying out remediation activities on remediated nitrate salt (RNS) and unremediated nitrate salt (UNS) waste containers at Los Alamos National Laboratory (LANL). The assessment includes a review of the waste streams consisting of 60 RNS, 29 aboveground UNS, and 79 candidate belowground UNS containers that may need remediation. The waste stream characteristics were examined along with the proposed treatment options identified in the Options Assessment Report . Two primary approaches were identified in the five candidate treatment options discussed in the Options Assessment Report: zeolite blending and cementation. Systems that could be used at LANL were examined for housing processing operations to remediate the RNS and UNS containers and for their viability to provide repackaging support for remaining LANL legacy waste.

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DOE OSTI2016

Corrosion Testing of Carbon Steel Exposed to Sludge Heel Chemical Cleaning Solutions - 16109

As efforts continue to treat and dispose of millions of gallons of legacy radioactive materials from the production of nuclear weapons, non-compliant waste storage tanks will gradually be emptied of the bulk waste volume leaving heel materials requiring removal prior to tank closure. The waste heel slurries are distributed on the floor of large, million gallon tanks, which frequently contain numerous obstructions that limit the effectiveness of mechanical removal methods. As a result, chemical cleaning methods are needed for the effective removal of the heels as well as chemical scales that are present on the tank walls and other interior surfaces. Two tank cleaning technologies have already been implemented at the Savannah River Site (SRS): Low Temperature Aluminum Dissolution (LTAD) and Bulk Oxalic Acid Cleaning (BOAC). Recent chemical cleaning efforts on SRS Tank 12 were very successful with regard to bulk sludge heel (especially for Al, Fe, and U phases) and beta/gamma radionuclide removal. Although chemical cleaning using these technologies has been shown to be effective, no disposition path has been identified for oxalate added during BOAC, and insoluble oxalate salts are accumulating within the SRS tank farm and waste processing facilities (evaporators, etc.). Extensive sludge washing is also required to remove moderately soluble sodium oxalate salts prior to sludge vitrification in the SRS Defense Waste Processing Facility (DWPF). As a result, oxalate additions to the tank farm need to be minimized by the use of supplementary acids to assist sludge removal in OA or the use of other cleaning reagents or processing strategies. Previous SRNL testing revealed the importance of pH control for BOAC and recommended the use of a supplementary acid (dilute HNO{sub 3}) with OA to minimize oxalate additions. Primary drivers in SRS Tank Closure Performance Assessments, which evaluate the fate and impact of tank sludge residuals on a geological timescale, are the removal of alpha emitting radionuclides present at low concentrations such as Pu, Am, and Np, which are not highly soluble in currently utilized chemical cleaning reagents. Scoping studies conducted at SRNL revealed promising methods to dissolve the actinides within the HLW tank heels. Oxidation of the actinides with permanganate in either strong caustic (10 M NaOH) or dilute acidic (0.2 M HNO{sub 3}) solutions was shown to result in dissolution of oxy/hydroxide phases of these metals in the absence of major sludge phases. The waste tanks at SRS and Hanford are constructed of carbon steel. Carbon steel corrodes rapidly in most acidic and/or oxidizing environments. Previous electrochemical corrosion testing indicated that the corrosion rates in dilute HNO{sub 3} with OA may result in manageable corrosion rates, over short exposure periods. Additional corrosion testing was performed to provide data for the nitric acid/OA blend as well as the acid/permanganate and caustic permanganate based chemical cleaning solutions. Corrosion rates from the passive coupon tests for the nitric acid/oxalic acid blend were significantly greater than those observed previously with the electrochemical testing. This result may be due to simultaneous electrochemical and chemical dissolution of the steel. The corrosion rate data from the coupon tests in the nitric acid/oxalic acid blend were evaluated to determine the degree of potential structural damage. The corrosion rates, although relatively high, would not be expected to cause damage that would reduce the capacity of the tank primary wall if the process is completed within a month. The corrosivity of sodium permanganate in four proposed cleaning solutions, 1 nitric acid solution and 3 sodium hydroxide solutions, was studied by electrochemical methods at room temperature. The corrosion rates were significantly less aggressive than in the nitric acid/oxalic acid blend corrosion tests. Passive coupon tests needs to be performed in these solutions to ensure that chemical dissolution of the metal is not occurring simultaneously. (authors)

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DOE OSTI2015

Support of SRS salt-stone regulatory compliance - 15586

The SRS salt-stone Facility is designed and permitted to immobilize and dispose of low-level radioactive and hazardous liquid waste (salt solution) remaining from the processing of radioactive material at the Savannah River Site. The salt-stone Processing Facility (SPF) is permitted as a wastewater treatment facility per the South Carolina Department of Health and Environmental Control Regulations (SCDHEC) South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-67. The salt-stone Disposal Facility is permitted as a solid waste landfill site, as defined by SCDHEC Regulations R61-107.19. The salt-stone Sampling and Analysis Plan provides the SCDHEC with the chemical and physical characterization strategy for the salt solution and salt-stone waste form. During operation, samples are collected from Tank 50H on a quarterly basis and are analyzed by the Savannah River National Laboratory (SRNL) for over chemical and radioactive components. The analytical results are compared to the salt-stone waste acceptance criteria in a technical report. SRNL also prepares salt-stone samples from the quarterly sample and premix binder materials provided by the SPF to determine the non-hazardous nature of the grout to meet the requirements of the SCDHEC regulation for the 'Characteristic of Toxicity', R.61-79.261.24(b) and the 'Universal Treatment Standards', R.61-79.268.48.4. salt-stone samples are collected, size reduced, packaged, and shipped to an independent United States EPA certified laboratory, or a laboratory that holds an equivalent certification such as one from the National Environmental Laboratory Accreditation Program to perform the EPA Toxicity Characteristic Leaching Procedure (TCLP) extraction of the salt-stone sample. SRNL reviews the results of the TCLP analysis and prepares a report detailing the sample preparation and analysis that is used by the salt-stone Facility to demonstrate the nonhazardous nature of the salt-stone waste form. (authors)

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DOE OSTI2021

Tank 48H Tetraphenylborate Mitigation: Simulant Studies using Sodium Permanganate

Tank 48H currently holds legacy material containing organic tetraphenylborate (TPB) compounds from the operation of the In-Tank Precipitation (ITP) process. TPB was added during the ITP process to precipitate the otherwise soluble cesium as insoluble cesium TPB (CsTPB), but excessive benzene generation from TPB degradation curtailed this treatment method. The contents of Tank 48H, which include approximately 26.000 kg of potassium TPB (KTPB) and trace CsTPB, are not compatible with the waste treatment facilities at the Savannah River Site (SRS) since the organic content and the associated flammability issues pose a challenge to the salt processing and sludge processing facilities within the liquid waste system. An in-tank process to remove (or decompose) TPB safely would be of great value. Previous testing at Savannah River National Laboratory (SRNL) demonstrated the destruction of glycolate via chemical oxidation using sodium permanganate with simulated and radioactive waste. Scoping tests were performed to study the destruction of TPB to determine if the contents of Tank 48H would be amenable to the same type of destruction. Partial destruction of TPB was observed in Tank 48H simulants under mild conditions (e g., pH 11, room temperature) with no definitive indication of benzene generation. To build upon the success of the scoping tests, an additional study was requested to provide a better understanding of the underlying chemistry for Tank 48H content destruction using sodium permanganate. Three experiments were performed with Tank 48H simulants at 40 °C to determine the efficacy of using sodium permanganate for TPB destruction. Three starting pH values were selected: 1) pH 11 for comparison with the previous work at room temperature, 2) pH 10 as the minimum pH recommended by the Corrosion Control Program (CCP) for in-tank processing, and 3) pH 8 to determine the effectiveness of TPB destruction at near neutral pH. While below the allowable pH for the CCP, the experiment at pH 8 was performed to study the TPB-Permanganate reaction under more extreme conditions and further verify the potential for out-of-tank processing.

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DOE OSTI2016

Mercury Phase II Study - Mercury Behavior in Salt Processing Flowsheet

Mercury (Hg) in the Savannah River Site Liquid Waste System (LWS) originated from decades of canyon processing where it was used as a catalyst for dissolving the aluminum cladding of reactor fuel. Approximately 60 metric tons of mercury is currently present throughout the LWS. Mercury has long been a consideration in the LWS, from both hazard and processing perspectives. In February 2015, a Mercury Program Team was established at the request of the Department of Energy to develop a comprehensive action plan for long-term management and removal of mercury. Evaluation was focused in two Phases. Phase I activities assessed the Liquid Waste inventory and chemical processing behavior using a system-by-system review methodology, and determined the speciation of the different mercury forms (Hg+, Hg++, elemental Hg, organomercury, and soluble versus insoluble mercury) within the LWS. Phase II activities are building on the Phase I activities, and results of the LWS flowsheet evaluations will be summarized in three reports: Mercury Behavior in the Salt Processing Flowsheet (i.e. this report); Mercury Behavior in the Defense Waste Processing Facility (DWPF) Flowsheet; and Mercury behavior in the Tank Farm Flowsheet (Evaporator Operations). The evaluation of the mercury behavior in the salt processing flowsheet indicates, inter alia, the following: (1) In the assembled Salt Batches 7, 8 and 9 in Tank 21, the total mercury is mostly soluble with methylmercury (MHg) contributing over 50% of the total mercury. Based on the analyses of samples from 2H Evaporator feed and drop tanks (Tanks 38/43), the source of MHg in Salt Batches 7, 8 and 9 can be attributed to the 2H evaporator concentrate used in assembling the salt batches. The 2H Evaporator is used to evaporate DWPF recycle water. (2) Comparison of data between Tank 21/49, Salt Solution Feed Tank (SSFT), Decontaminated Salt Solution Hold Tank (DSSHT), and Tank 50 samples suggests that the total mercury as well as speciated forms in the assembled salt batches in Tanks 21/49 pass through the Actinide Removal Process (ARP) / Modular Caustic Side Solvent Extraction Unit (MCU) process to Tank 50 with no significant change in the mercury chemistry. (3) In Tank 50, Decontaminated Salt Solution (DSS) from ARP/MCU is the major contributor to the total mercury including MHg. (4) Speciation analyses of TCLP leached solutions of the grout samples prepared from Tank 21, as well as Tank 50 samples, show the majority of the mercury released in the solution is MHg.

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DOE OSTI2021

Road Map for Developing Iron Phosphate Waste Forms for Salt Wastes

In this report, issues that must be addressed to advance the technology readiness level of phosphate glass waste forms being developed to immobilize high-level radioactive salt waste streams are identified, the states of understanding various technical aspects of formulation, processing, and performance are summarized, and approaches supporting further development are recommended. Processing results in dehalogenation of the waste salt, capture of the gaseous halide-bearing species, and immobilization of the residual salt components in a phosphate glass waste form. The approach is suitable for high-level salt waste from electrochemical reprocessing and molten salt reactors. The technology has been demonstrated for chloride-based salts and may also be suitable for the treatment and immobilization of fluoride-based and iodide-bearing waste salts. Aspects of the process requiring further development are identified and approaches recommended.

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DOE OSTI2016

Mercury Phase II Study - Mercury Behavior across the High-Level Waste Evaporator System

The Mercury Program team’s effort continues to develop more fundamental information concerning mercury behavior across the liquid waste facilities and unit operations. Previously, the team examined the mercury chemistry across salt processing, including the Actinide Removal Process/Modular Caustic Side Solvent Extraction Unit (ARP/MCU), and the Defense Waste Processing Facility (DWPF) flowsheets. This report documents the data and understanding of mercury across the high level waste 2H and 3H evaporator systems.

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DOE OSTI2020

TCCR Operational Summary and Optimization for Tank 9 Processing - 20314

Savannah River Remediation (SRR) manages and operates the liquid waste facilities at Savannah River Site (SRS) for the Department of Energy (DOE). Stored liquid waste is a complex mixture of insoluble solids (sludge) and soluble salts in an alkaline solution. SRR has deployed the Tank Closure Cesium Removal (TCCR) system, a tank-side ion exchange process, to remove radioactive cesium from salt waste and enable onsite disposal of the resulting decontaminated salt solution as low-level waste at the Saltstone facilities. The TCCR system consists of two prefilters, four ion exchange (IX) columns, one resin trap, and a ventilation system. The IX process uses a form of inorganic crystalline silicotitanate (CST), which has a high affinity for cesium and other alkali metals, strontium, and actinides. This process is currently deployed utilizing salt feed from Tank 10, with future plans to dissolve solid salt in Tank 9 and transfer the salt solution to Tank 10 for processing through TCCR. The feed for TCCR must be created from salt-cake in Tank 10 through a dissolution process. Once enough salt has been dissolved, a qualification process is entered. This process characterizes the feed and ensures the cesium loading on the columns will not cause boiling of waste within the columns during or after processing. Once the batch has been qualified, salt waste is fed to the TCCR system through a transfer pump in the center of the tank. The waste is filtered through a set of two shielded, dead-end prefilters that prevent solids buildup in the columns. The filtered salt solution then travels to the shielded IX columns, which can be operated individually or in series, where the cesium is sorbed on the CST media. The decontaminated salt solution (DSS) then travels through a resin trap and out of the module to Tank 11. TCCR has successfully processed approximately 795,000 L of Tank 10H radioactive salt waste over two batches to date. There has not yet been a system induced shutdown. The prefilters performed as expected with only minor degradation in recovery of differential pressure after a backflush sequence. The time between backflushes decreased as each batch reached the end of processing. The hydraulics in the IXCs mostly performed as expected at all flow rates, except for one IXC that will be further investigated during Batch 3 processing. The TCCR system has shown some opportunities for more efficient processing during the length of the demonstration so far. For future processing of material from Tank 9H through Tank 10H and the TCCR unit, TCCR 1A will implement changes to the prefilters and the IXCs. The prefilters will have an increased surface area and a new filter media in an effort to increase time between filter swaps and improve backwashing cleaning capability. The IXCs will have a reduced diameter to allow for increased heat transfer out of the column and increased loading of Cs-137. Additionally, a new form of CST with an increased kinetic performance is being investigated for use during TCCR 1A operation. (authors)

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DOE OSTI2021

Deliquescence of Eutectic LiCl-KCl Diluted with NaCl for Interim Waste Salt Storage

Molten eutectic LiCl-KCl salt is a widely used electrolyte for electrorefining uranium from spent nuclear fuel. Due to the hygroscopic nature of this salt, such operations must be performed under controlled atmospheric conditions, and waste salts require careful storage to avoid deliquescence and corrosion of container materials. This study investigated a potential processing path for reducing the degree of deliquescence through dilution to varying extents with NaCl. Here, the hydration behavior of LiCl-KCl salts diluted with NaCl was evaluated in terms of mass gain due to water absorption, degree of deliquescence (including first appearances of standing water), and evidence of corrosion to stainless steel containers in a humid air environment (40°C, 20% relative humidity). In this humid air environment, pure eutectic LiCl-KCl exhibited a 50 mass % increase due to water absorption and showed evidence of standing water after 24 h. Waste salt diluted with NaCl required loadings of 89 mass % NaCl in order to prevent deliquescence and exhibited a 3 mass % increase due to water absorption. After periodic observation for 48 h, standing water was observed near all ingots with the exception of the 89 mass % NaCl samples. Dilution with 89% NaCl was also found to reduce evidence of corrosion when stored in stainless steel crucibles. While dilution with NaCl greatly decreases steady-state hydration, the storage volume is increased ~10× through this procedure.

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DOE OSTI2016

Evaluation of no-MST operations in the SRS ARP for Hanford LAWPS

The Savannah River Site (SRS) Actinide Removal Process has been processing salt waste since 2008. This process includes a filtration step in the 512-S facility. Initial operations included the addition, or strike, of monosodium titanate (MST) to remove soluble actinides and strontium. The added MST and any entrained sludge solids were then separated from the supernate by cross flow filtration. During this time, the filter operations have, on many occasions, been the bottleneck process limiting the rate of salt processing. Recently, 512-S- has started operations utilizing “No-MST” where the MST actinide removal strike was not performed and the supernate was simply pre-filtered prior to Cs removal processing. Direct filtration of decanted tank supernate, as demonstrated in 512-S, is the proposed method of operation for the Hanford Low Activity Waste Pretreatment System (LAWPS) facility. Processing decanted supernate without MST solids has been demonstrated for cross flow filtration to provide a significant improvement in production with the SRS Salt Batches 8 and 9 feed chemistries. The average filtration rate for the first 512-S batch processing cycle using No-MST has increased filtrate production by over 35% of the historical average. The increase was sustained for more than double the amount of filtrate batches processed before cleaning of the filter was necessary. While there are differences in the design of the 512-S and Hanford filter systems, the 512-S system should provide a reasonable indication of LAWPS filter performance with similar feed properties. Based on the data from the 512-S facility and with favorable feed properties, the LAWPS filter, as currently sized at over twice the size of the 512-S filter (532 square feet filtration area versus 235 square feet), has the potential to provide sustained filtrate production at the upper range of the planned LAWPS production rate of 17 gpm.

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DOE OSTI2018

Coupled Processes Modeling in Rock Salt and Crushed Salt Including Halite Solubility Constraints: Application to Disposal of Heat-Generating Nuclear Waste

This work presents numerical modeling of coupled thermal, hydraulic and mechanical processes in rock salt and crushed salt considering halite solubility constraints. The TOUGH-FLAC simulator is used, with a recently enhanced Equation-Of-State module that includes the thermodynamic properties of aqueous fluids of variable salinity. Laboratory and field scale tests performed on rock salt and crushed salt under temperature gradients are modeled first to evaluate the capabilities of the simulator to reproduce important features, such as porosity changes induced by halite dissolution/precipitation, and brine and heat migration. Since the results are quite satisfactory, the simulator is used to predict the long-term response of a generic salt repository for heat-generating nuclear waste. To evaluate the impacts of halite solubility on the predictions, two simulations that respectively consider or neglect solubility constraints are performed. In the scenario studied, the results are not significantly affected by dissolution/precipitation, and only some differences are observed due to changes in porosity, but the dominating processes remain the same. Finally, with the new provisions, TOUGH-FLAC is more complete in terms of processes occurring around a heat-releasing nuclear waste package and can therefore provide more accurate predictions of the long-term performance of a nuclear waste repository in salt formations.

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DOE OSTI2020

Issue Resolution During the Development of the Performance Assessment for the Savannah River Site Saltstone Disposal Facility - 20125

In 2019, Savannah River Remediation developed a revision to the performance assessment (PA) on behalf of the U.S. Department of Energy (DOE) Savannah River Operations Office (SR) for the near-surface disposal of low-level waste at the Savannah River Site (SRS) Saltstone Disposal Facility (SDF). Soluble waste from SRS Tank Farms undergoes salt processing to remove cesium and other high-activity constituents. The low-activity decontaminated salt solution (DSS) is then immobilized by mixing it into a cementitious waste form known as saltstone. After mixing, the saltstone is poured into leak-tight concrete vaults, known as saltstone disposal units (SDUs), where the waste form cures. By the time of facility closure, the SDF is expected to consist of 15 SDUs with a combined capacity of 1.06 E+09 L (280 Mgal) of cured saltstone. The facility operates under a Disposal Authorization Statement from DOE and a permit from the South Carolina Department of Health and Environmental Control (SCDHEC). Since the start of operations in 1990, the SDF has received almost 6.7 E+07 L (18 Mgal) of DSS, resulting in the safe disposal of 2.7 E+16 Bq (7.3 E+05 Ci) of activity. Due to the radioactive decay of short-lived contaminants, the total remaining activity in the disposed waste is estimated to be approximately 1.4 E+16 Bq (3.9 E+05 Ci), as of September 2018. The Disposal Authorization Statement requires a demonstration that the system of engineered and natural features of the disposal facility will limit releases from the facility and be protective of human health and the environment for at least the next 1,000 years. The long-term performance of the facility was evaluated under the requirements of the DoE's Radioactive Waste Management Manual (US DOE Manual 435.1-1). Simulations were performed to demonstrate that the disposal facility would meet performance objectives specified in the manual. The evaluation was based on numerical models that simulate the releases of contaminants from the saltstone waste form. Contaminants were transported through groundwater and air pathways to points of assessment to evaluate compliance (i.e., 100 m from the SDUs). In addition, the potential consequences of an inadvertent human intrusion (IHI) were also evaluated. A number of issues were overcome during the development of these simulations. These issues were identified as part of internal technical reviews. Simulations are developed by people and people make mistakes, so the internal technical review process is a vital step in PA development. Specific examples of resolved issues include a unit-conversion error, an inappropriate definition for a model boundary condition, a model time-stepping issue, and an error in the calculation for the buildup of contaminants in soil. Actions taken to address these issues resulted in an improved product with a better supported technical basis and more defensible results. The identification and correction of these issues are discussed. By understanding these issues, model developers and technical reviewers working on PAs in the future may avoid repeating these types of mistakes. Transparency with respect to these mistakes builds trust between waste management sites, regulators, and stakeholders. (authors)

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DOE OSTI2020

Update to the Performance Assessment for the Savannah River Site Saltstone Disposal Facility - 20124

In 2019, Savannah River Remediation developed a revision to the performance assessment (PA) on behalf of the U.S. Department of Energy (DOE) Savannah River Operations Office (SR) for the near-surface disposal of low-level waste at the Savannah River Site (SRS) Saltstone Disposal Facility (SDF). Soluble waste from SRS Tank Farms undergoes salt processing to remove cesium and other high activity constituents. The low-activity decontaminated salt solution (DSS) is then immobilized by mixing it into a cementitious waste form known as saltstone. After mixing, the saltstone is poured into leak-tight concrete vaults, known as saltstone disposal units (SDUs), where the waste form cures. By the time of facility closure, the SDF is expected to consist of 15 SDUs with a combined capacity of 1.06 E+09 L (280 Mgal) of cured saltstone. The facility operates under a Disposal Authorization Statement (DAS) from DOE and a permit from the South Carolina Department of Health and Environmental Control (SCDHEC). Since the start of operations in 1990, the SDF has received almost 6.7 E+07 L (18 Mgal) of DSS, resulting in the safe disposal of 2.7 E+16 Bq (7.3 E+05 Ci) of activity. Due to the radioactive decay of short-lived contaminants, the total remaining activity in the disposed waste is estimated to be approximately 1.4 E+16 Bq (3.9 E+05 Ci), as of September 2018. The Disposal Authorization Statement requires a demonstration that the system of engineered and natural features of the disposal facility will limit releases from the facility and be protective of human health and the environment for at least the next 1,000 years. The long-term performance of the facility was evaluated under the requirements of the DoE's Radioactive Waste Management Manual (US DOE Manual 435.1-1). Simulations were performed to demonstrate that the disposal facility would meet performance objectives specified in the manual. The evaluation was based on numerical models that simulate the releases of contaminants from the saltstone waste form. Contaminants were transported through groundwater and air pathways to points of assessment to evaluate compliance. In addition, the potential consequences of an inadvertent human intrusion (IHI) were also evaluated. In accordance with the guidance and recommendations in US DoE's technical standard for DAS, deterministic and probabilistic analyses were performed to demonstrate the SDF system, which includes the engineered cover system, the SDUs, the waste form, and the natural features of the site, provides a reasonable expectation that saltstone disposal will meet performance objectives. Post-closure doses to future members of the public (MOP) were evaluated; dose estimates from the air pathway were well below the 1.0 E-04 Sv/yr (10 mrem/yr) performance objective within 1,000 years, and the calculated doses to the MOP from all exposure pathways, including the groundwater and air pathways, were well below the 2.5 E-04 Sv/yr (25 mrem/yr) performance objective within 1,000 years. Doses following an assumed intrusion event within the facility boundaries were well below the IHI performance objectives, where the acute IHI dose was below the acute IHI dose performance objective of 5.0 E-03 Sv (500 mrem) and the chronic IHI dose was below the chronic IHI dose performance objective of 1.0 E-03 Sv/yr (100 mrem/yr) within 1,000 years. (author)

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DOE OSTI2018

Crystalline Silicotitanate ion exchange media long-term storage evaluation

Crystalline Silicotitanate a (CST) ion exchange media is produced in an engineered form, known as IONSIV™ IE-911. This material has been tested extensively for use in removing 137 Cs from Savannah River Site (SRS) and Hanford site tank waste. This inorganic material is highly selective for removing cesium ions, even in the presence of high concentrations of sodium and potassium ions and is tolerant of the harsh conditions of the varying tank chemistries. A batch of caustic washed IE-911 that had been produced by the supplier and received at SRS in 2001 was tested to verify that it was still in good condition after the extended storage. This testing involved standard batch contacts to measure the removal of Cs from simulated tank waste. This report describes the results of testing performed to verify that the IE-911 matches the performance of previous batches of this media. To confirm that the performance of IE-911 is similar to previous batches and that no degradation had occurred during storage, batch contact testing was performed using SRNL’s standard methodology using a non-radioactive SRS Average Waste Simulant composition utilized in previous studies. Batch contact kinetics testing was also conducted to determine cesium loading on this batch versus time. The cesium equilibrium loading isotherm in SRS simulant at 23°C was compared to ZAM isotherm model predictions and previous test results with IE-911 media. The measured and predicted distribution coefficients and cesium loading (“Q”) are shown in Table ES-1. The physical properties of the material were also determined, including density, moisture content, and particle size.

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DOE OSTI2016

DWPF Safely Dispositioning Liquid Waste

The only operating radioactive waste glassification plant in the nation, the Defense Waste Processing Facility (DWPF) converts the liquid radioactive waste currently stored at the Savannah River Site (SRS) into a solid glass form suitable for long-term storage and disposal. Scientists have long considered this glassification process, called “vitrification,” as the preferred option for treating liquid radioactive waste.

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DOE OSTI2015

Engineering scale demonstration of a prospective cast Stone process - 15584

A cementitious waste form is one of the alternatives being evaluated for Supplemental Immobilization of Hanford Low Activity Waste (LAW). Washington River Protection Solutions (WRPS) is collecting data on cementitious or pozzolanic waste forms such as Cast Stone. The goal of this project for WRPS is to obtain data on the performance of the Cast Stone waste form for immobilizing LAW. As part of the data package being assembled, an engineering-scale demonstration with non-radioactive simulants was performed at the Savannah River National Laboratory (SRNL) using the Scaled Continuous Processing Facility (SCPF) to fill a container with simulated Cast Stone grout. Legacy salt solution from previous Hanford salt waste testing was trimmed to resemble the average composition generated from the Hanford Tank Waste Operation Simulator (HTWOS) used in the screening tests. The dry blend materials, ordinary portland cement (OPC), Class F fly ash, and ground granulated blast furnace slag (GGBFS or BFS) were obtained from Lafarge North America in Pasco, WA. Over three days, the SCPF was used to fill a 6,056 liter (1600 gallon) container with simulated Cast Stone grout. The tank was instrumented with x-, y-, and z-axis thermocouples to monitor curing temperature and two formed core sampling vials. The target production rate was 5.7 L/min (1.5 gpm). This required a salt solution flow rate of approximately 3.78 L/min (1 gpm) and a premix feed rate of approximately 263 kh/h (580 lb/h). The final surface slope at a fill height of 100 cm was 2.5-3.8 cm across the 260 cm (8.5 ft) diameter tank. During processing, grout was collected from both the mixer discharge and the discharge into the tank. These samples were stored in a humid environment either in a closed box proximal to the tank or inside the laboratory. (authors)

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